Field of the Invention
The invention relates to a fuel element for a pressurized water reactor, with a laterally open skeleton containing control-rod guide tubes, to which are fastened a plurality of spacers and also a fuel element head and a fuel element foot. Gastight multilayer cladding tubes are inserted into the skeleton and in each case surround a column of fuel pellets. The invention relates, furthermore, to a method for producing the cladding tubes.
FIG. 1 illustrates a fuel element of this type, with a head 1, a foot 2 and spacers 3 and 4 that are fastened to guide tubes 5, thus producing a laterally open skeleton, into which fuel rods 6 are inserted. During operation, cooling water flows from the bottom upward through the fuel element and can also enter adjacent fuel elements laterally from the interspaces between the fuel rods. It can also be seen from FIG. 1 that additional mixing grids 7, which serve as carriers for flow guide blades, may be provided between the spacers 4 in an upper part of the fuel element. Such flow guide blades are advantageously likewise provided in the upper part of the fuel element, at least on a top side of the spacers 4, in order to achieve a turbulent mixing of the cooling water and a better flow onto the fuel rods 6.
Corresponding blades are described, for example, in Published, Non-Prosecuted German Patent Application DE 15 64 697 A and are reproduced in FIG. 2. It is also possible, however, to have other spacers (for example, formed of sleeves welded to one another) and other geometries of flow guide blades 8, while a different number of mixing vanes may also be provided in the interspaces between adjacent fuel rods 6.
In the pressurized water reactor, only a small fraction (normally at most 5%) of the liquid cooling water is evaporated on the outer surfaces of the fuel rods, on the contrary the heat generated in the fuel by nuclear fission is discharged essentially in that water having a corresponding temperature and maintained under high pressure is transported away by convection. In contrast, a boiling water reactor operates with lower pressure and lower temperatures, the heat from the fuel rods being transported away, at least in the upper part of the fuel element, essentially by isothermal evaporation in a two-phase mixture. In this case, it is necessary to channel the steam that occurs. The fuel elements are therefore surrounded laterally by fuel element boxes.
The techniques of the boiling water reactor and of the pressurized water reactor have developed in different directions. For the purpose of plant protection and for similar reasons, the pressurized water has admixed with it, for example, lithium hydroxide and similar additives which cannot be used in boiling water and lead to a different water chemistry (for example, a different oxygen concentration). The size and number of fuel rods in the fuel elements and the configuration of control elements in the reactor core are also different. The differences in the temperature and pressure of the cooling water also lead to different loads on the cladding tubes and to a different behavior of the fuel, in particular to different time constants of the reactor core when the latter is considered as a self-contained control system with feedback.
The result of this different control behavior is that the power output of pressurized water reactors is changed only very slowly, that is to say the pressurized water reactor is operated almost virtually in the steady state and is suitable particularly for covering basic loads. For covering peak loads of the consumer connected to the reactor, boiling water reactors, the power output of which is run up, for example, substantially more quickly and in a ramp-like manner, are more suitable. The result of this is that the cladding tubes, which are already exposed on their outer surface, according to the different water chemistry and operating temperatures, to different chemical loads (for example, nodular corrosion in the boiling water reactor or uniform corrosion in the pressurized water reactor) and have to withstand different operating pressures, are also subjected to different loads on the inside.
The outcome of this has been that the cladding tubes of boiling water reactors are formed, as a rule, of a different alloy (to be precise, zircaloy-2) from the cladding tubes of pressurized water reactors for which zircaloy-4 was developed. A zirconium alloy with 2.5% niobium, which is also used in Russian light-water cooled reactors, is also known for the pressure tubes of high-temperature reactors.
Table 1 indicates the standardized composition of industrially pure zirconium for the nuclear industry (so-called “zirconium sponge”), zircaloy-2 (“zry-2”), zircaloy-4 (“zry-4”) and zirconium niobium (“Zr/Nb”), oxygen being considered as an impurity acceptable in small quantities, even when, because of its hardening effect on zirconium, it is often desirable and is therefore added.
If use is made of a higher enrichment of the fuel pellets with fissionable isotopes of uranium and/or plutonium and therefore of a greater useful energy content (so-called “burn-up”) of the fuel, then the fuel elements can remain in the core for longer, should their cladding tubes be capable of meeting the corresponding requirements due to the longer service life. Therefore, in pressurized water fuel elements, the outer surfaces of the cladding tubes must be particularly resistant to the uniform corrosion occurring in the pressurized water and should not be pressed onto the fuel by the increased pressure, even in the event of relatively long service lives, in such a way that they thereby experience damage. In the development of cladding tubes that meet the increased requirements of a longer service life in the pressurized water reactor, it is therefore necessary to pay particular attention to the mechanical stability of the entire tube and to the resistance of the outer surface to uniform corrosion.
These conditions are fulfilled satisfactorily by single-layer cladding tubes, such as are described in European Patent EP 0 498 259 B and, in general, consist of zirconium with 0.8 . . . 1.7% Sn, 0.07 . . . 0.5% Fe, 0.05 . . . 0.35% Cr, 0.07 . . . 0.2% O, up to about 0.015% Si and up to a maximum of 0.1% Ni. In this context, it has proved particularly important that the metals, Fe, Cr and Ni, which are virtually insoluble in zirconium and are precipitated (so-called “secondary precipitations”) as intermetallic compounds (“secondary phases”), have an average particle size of about 0.1 to 0.3μ. The particle size is set by the thermal treatment to which the alloy is subjected after it has first been brought to a temperature at which the precipitations are dissolved (so-called “solution annealing”) and has then been rapidly cooled (so-called “quenching”). The resulting size and distribution of the secondary precipitations can be calculated by a “particle growth parameter” and in manufacturing practice are set by a cumulative “standardized annealing duration” AA=Σti·exp(−Q/T),in which T is the temperature in Kelvin during a manufacturing step i, ti is the duration of the manufacturing step and Q corresponds to an activating energy, and the value Q=40,000 Kelvin may be adopted.
FIG. 3 shows the daily growth of the uniform oxidation layer on the surface of a cladding tube formed of zircaloy-4 in a pressurized water reactor at operating temperatures of about 300° C. as a function of the standardized annealing duration A which was used in the production of the cladding tube. In general, for pressurized water reactors, standardized annealing durations of between 2·10−18 and 50·10−18 hours are considered favorable for zircaloy-like alloys of this type, such as are described in European Patent EP 0 498 259 B (Garzarolli et al. in “Zirconium in the Nuclear Industry: Eighth International Symposium”, Philadelphia 1989 (ASTM Special Technical Publication 1023), pages 202 to 212). However, such a high annealing duration conflicts with the efforts of a person skilled in the art, by a pilgrim-step method with cold formings, to break down the alloy grain, which likewise ripens into large grains at high temperatures, into small grains by cold formings, in order to increase the mechanical stability of the cladding tube, since a fine grain leads to high stability along with high ductility. Consequently, according to the patent specification mentioned, the high standardized annealing duration is achieved by the quenched material first being forged, still at a high temperature, before it is extruded to form a tube blank and is then cold-formed in subsequent pilgrim steps with moderate intermediate annealings.
Another way is to have a composite tube that, as a so-called “duplex”, formed of a relatively thick matrix layer with a thin outer protective layer formed of another zirconium alloy. The matrix ensures the necessary mechanical stability, while the outer protective layer is resistant to the uniform corrosion posing a threat in the pressurized water reactor. Such a duplex is described for the first time in European Patent EP 0 212 351 B, where 0.1 to 1% V and up to 1% Fe is used as alloying additives for the outer protective layer. European Patent EP 0 301 295 B describes a duplex, in which the outer alloy contains 0.2 to 3% Nb and/or a total content of Fe, Cr, Ni and Sn of between 0.4 and 1% (remainder: in each case zirconium of industrial purity). It is known from European Patent EP 0 630 514 B that an outer layer of this type for a zircaloy matrix may also contain a larger total content of Fe, Cr, Ni, Sn, insofar as specific restrictions are maintained for the individual alloying additives, in particular the tin content is below the tin content of the zircaloy. The cladding tubes mentioned have proved appropriate, even under the operating conditions of the pressurized water reactor, and make it possible to achieve the desired long service lives.
The graph of FIG. 3 would be entirely different in the case of a boiling water reactor. There, because of the lower operating temperatures, virtually no uniform corrosion occurs, but oxide pustules are formed. Here, high secondary precipitations cannot act as any of the pustules that, however, are avoided when the material of the secondary phases is finely distributed and has undergone only a particularly low standardized annealing duration. Often, however, cladding tubes of boiling water fuel rods exhibited corrosion damage that emanated from inside the tubes and was attributed to stress crack corrosion. Such damage was minimized by a composite tube, in which a matrix of zircaloy had on the inside a protective layer of industrially pure zirconium, that is to say a soft material, but one susceptible to corrosion. In this case, however, the susceptibility of pure zirconium to corrosion is a disadvantage, since the situation is unavoidable where, in rare instances, due to slight damage in the tube, water from the boiling water reactor enters the cladding tube interior and then triggers corrosion leading to extensive cracks by which the water of the reactor may be contaminated to a substantially greater extent than by a multiplicity of fuel rods with locally limited damage. Instead of a protective layer of pure zirconium, therefore, a protective layer is often used, in which the zirconium contains up to 1% of another alloying additive. Thus, European Patent EP 0 726 966 B describes a cladding tube with a thick matrix layer of zircaloy, in which the secondary precipitations have a particle size of between about 0.03 and 0.1μ, and a lining of zirconium with 0.2 to 0.8% iron is bonded metallurgically to the inside.
The composite tube is particularly advantageous in the boiling water reactor, because, due to the small size of the secondary precipitations on the outer surface, a particularly low A-value becomes necessary, which, in the case of the appropriate alloying of the protective layer on the inside of the cladding tube, likewise brings about only a slight growth of secondary precipitations and grain, so that the inside is both protected more effectively against corrosion and remains soft because it is not subject to any excessive dispersion hardening as a result of Fe secondary precipitations.
However, a cladding tube of this type, configured for boiling water conditions, is entirely unsuitable for pressurized water applications, since the small size of the secondary precipitations on the outer surface would accelerate the uniform corrosion and necessitate an exchange of the cladding tube even after short service lives. On the other hand, the inner lining is not necessary, even under the operating conditions of the pressurized water reactor which have existed hitherto, since, up to now, no damage emanating from the inner surface (stress crack corrosion) has been observed. Moreover, the power output of the pressurized water reactors is not changed rapidly in the ramp-like manner, as is customary in boiling water reactors. Instead, the control conditions of the pressurized water reactor make it necessary, in any case, for the power output to be changed only slowly, there being predetermined for the control a rate of change which also takes account of the fact that the cladding tubes are not to be subjected to inadmissible stress.
In the case of a higher enrichment of the fuel and longer service lives, even the behavior of the fuel itself must be taken into consideration. Since a multiplicity of gaseous fission products occur during decomposition, the fuel swells and thereby experiences an enlargement of volume which leads to a widening of the cladding tube, especially since the latter, in the course of time, particularly under the higher pressures of the pressurized water reactor, is compressed and creeps onto the fuel. When the fuel, which is in contact with the inside of the cladding tube even at a low reactor power output, is quickly heated as a result of a rapid increase in power output customary in the boiling water reactor, however, the thermal expansion of the fuel constitutes an additional load on the cladding tube. In configuration terms, the loads can be taken into account in as much as a gas collecting space is provided at least in the upper end of the fuel rods, a gap is left free between the cladding tubes and the fuel pellets and the fuel element is efficiently and quickly cooled, for example by the initially mentioned flow guide blades on the spacers and, if appropriate, additionally introduced intermediate grids. The load has hitherto been unimportant in the control of the power output of pressurized water reactors, since, in any case, in control terms a restricted rate of change of the power output seems permissible.